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DOE OSTI2022

Flash Recycling of Graphite Anodes

The ever-increasing production of commercial lithium-ion batteries (LIBs) will result in a staggering accumulation of waste when they reach their end of life. A closed-loop solution, with effective recycling of spent LIBs, will lessen both the environmental impacts and economic cost of their use. Presently, <5% of spent LIBs are recycled and the regeneration of graphite anodes has, unfortunately, been mostly overlooked despite the considerable cost of battery-grade graphite. Here, in this work, an ultrafast flash recycling method to regenerate the graphite anode is developed and valuable battery metal resources are recovered. Selective Joule heating is applied for only seconds to efficiently decompose the resistive impurities. The generated inorganic salts, including lithium, cobalt, nickel, and manganese, can be easily recollected from the flashed anode waste using diluted acid, specifically 0.1 M HCl. The flash-recycled anode preserves the graphite structure and is coated with a solid-electrolyte-interphase-derived carbon shell, contributing to high initial specific capacity, superior rate performance, and cycling stability, when compared to anode materials recycled using a high-temperature-calcination method. Life-cycle-analysis relative to current graphite production and recycling methods indicate that flash recycling can significantly reduce the total energy consumption and greenhouse gas emission while turning anode recycling into an economically advantageous process.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Mercury Speciation and Stabilization in Saltstone - 19103

Mercury is a contaminant of concern in Savannah River Site (SRS) high level waste (HLW). For five decades elemental mercury was used in the catalytic dissolution of aluminum cladding from the enriched-alloy uranium recovery process. Over 60,000 kg of mercury are now distributed in the SRS HLW tanks. The mercury speciation in the liquid waste streams resulting from conditioning HLW sludge for vitrification and in the decontaminated low activity sodium salt solution removed from the tanks was initially thought to be primarily ionic with a minor amount of elemental mercury. Recent sampling detected organic mercury, primarily methylmercury, HgCH{sub 3}{sup +} (MeHg{sup +}), and a trace amount of ethylmercury (EtHg) in these streams in addition to the ionic and elemental forms of mercury. The objective of this work was to identify the fate of the organic mercury in saltstone, a waste form containing portland cement, slag cement, and Class F fly ash. This information is needed to evaluate total mercury and organic mercury limits for the saltstone Waste Acceptance Criteria to assure regulatory compliance with mixed waste disposal requirements and to evaluate opportunities to enhance the waste form performance. Results to date indicate that the MeHg{sup +} reacts with sulfide in the cementitious reagents, slag in particular, to form nanoparticles of β-HgS, meta cinnabar, which is very insoluble and has a solubility product in water, pK{sub sp}, at 25 deg. C of 53. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Fundamental Electrochemical Properties of Liquid Metals in LiCl-KCl for Separation of Alkali/Alkaline-Earths (Cs, Sr, and Ba)

In electrorefiner, uranium is recovered from used nuclear fuel using an electrorefining process in which a metallic used nuclear fuel anode is oxidized into molten LiCl-KCl-UCl 3 electrolyte and pure U is preferentially reduced onto an inert cathode. While electrorefiner systems facilitate the recycling of substantial amounts of uranium from used nuclear fuel, they also contribute to the production of nuclear waste due to the build-up of fission products such as 90 Sr and 137 Cs in the molten salt electrolyte as they are electrochemically more active than U. The accumulation of alkali/alkaline-earth elements (Ba, Sr, Cs) in the electrolyte presents a problem as Sr and Cs isotopes have high heat densities and produce large amounts of highly ionizing radiation; these hazards combined with difficulty in removing the highly stable alkali/alkaline-earth elements from the electrolyte necessitates frequent replacement and disposal of the electrolyte, which then increases the overall volume of nuclear waste. This research project focuses on evaluating the viability of using liquid metal electrodes for electrochemical separation of alkali/alkaline-earths from molten salts utilizing the strong atomic interactions between candidate liquid metals and alkali/alkaline-earths. The strength of chemical interactions was quantified by determining the thermodynamic properties (e.g., activity) in liquid metals of Bi, Sb, and Pb. Thermodynamic properties, including activities, partial molar entropies, and partial molar enthalpies, were determined using electromotive force measurements for the Sr-Bi, Sr-Sb, Sr-Pb, Ba-Bi, and Ba-Sb binary systems in order to elucidate the strength of interactions between the alkaline-earth elements and each liquid metal, and to develop a comprehensive understanding of their behavior. Activities as low as a Sr = 10 -13 at x Sr = 0.04 at T = 888 K were measured as well as liquid state solubility as high as 40 mol% at 988 K in Sr-liquid metal systems; activities as low as aBa = 10 -15 at xBa = 0.05 at T = 888 K with liquid state solubility as high as 30 mol% at 988 K in Ba-liquid metal systems. Experimental data was used as input data towards computational efforts involving first-principles calculations as well as the CALPHAD (CALculation of PHAse Diagram) technique in the case of the Sr-Sb and Ba-Bi systems to develop improved phase diagrams and provide further basis for the use of computational models in elucidating strongly interacting binary systems. Attempts to remove Sr and Ba from molten salt electrolyte using an electrochemical cell with liquid metal cathodes were successful, with post-mortem elemental analysis of the electrodes confirming significant quantities of Sr (6.5 mol%) and Ba (12.8 mol%) deposited into Bi. Furthermore, deposition results correlated well with the deposition behavior predicted from the aforementioned electromotive force measurements, inviting the possibility of using liquid metal electrodes for selectively removing alkali/alkaline-earth elements from molten LiCl-KCl electrolyte to recycle the process salt in electrorefiner.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Uranyl oxalate species in high ionic strength environments: stability constants for aqueous and solid uranyl oxalate complexes

Uranyl ion, UO 2 2+ , and its aqueous complexes with organic and inorganic ligands can be the dominant species for uranium transport on the Earth surface or in a nuclear waste disposal system if an oxidizing condition is present. As an important biodegradation product, oxalate, C 2 O 4 2– , is ubiquitous in natural environments and is known for its ability to complex with the uranyl ion. Oxalate can also form solid phases with uranyl ion in certain environments thus limiting uranium migration. Therefore, the determination of stability constants for aqueous and solid uranyl oxalate complexes is important not only to the understanding of uranium mobility in natural environments, but also to the performance assessment of nuclear waste disposal. In this work, we developed a thermodynamic model for the UO 2 2+ –Na + –H + –Cl – –ClO 4 – –C 2 O 4 2– –NO 3 – –H 2 O system to ionic strength up to ~11 mol•kg –1 . We constrained the stability constants for UO 2 C 2 O 4 (aq) and UO 2 (C 2 O 4 ) 2 2– at infinite dilution based on our evaluation of the literature data over a wide range of ionic strengths up to ~11 mol•kg –1 . We also obtained the solubility constants at infinite dilution for solid uranyl oxalates, UO 2 C 2 O 4 •3H 2 O, based on the solubility data over a wide range of ionic strengths. The developed model will enable for the accurate stability assessment of oxalate complexes affecting uranium mobility under a wide range of conditions including those in deep geological repositories.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Nuclear Criticality Safety Assessment of Criticality Control Containers without Moderation Control at the Waste Isolation Pilot Plant

The Waste Isolation Pilot Plant (WIPP) provides for safe, permanent disposal of government-owned transuranic (TRU) and TRU mixed wastes. Receipt and disposal of waste at the WIPP site began in March 1999. The Sandia report, Consideration of Nuclear Criticality When Disposing of Transuranic Waste at the Waste Isolation Pilot Plant, addressed potential nuclear criticality safety issues based on the projected inventory characteristics known at the time [1]. As designs for inventory, waste forms, and disposal packages have changed, new analyses have been performed, and updates have been made to address any potential effects to the WIPP safety basis. New analyses performed include Saylor 2017 [2] and Brickner 2019 [3], which address certain waste containers with specified loadings under post-closure conditions. Both examined several hypothetical scenarios and included analyses to bound (from a criticality potential standpoint) credible configurations that could occur at WIPP during the repository regulatory post-closure disposal time period for feature, event, and process (FEP) considerations—10,000 years. During this post-closure period at WIPP, the screening of FEPs is governed by the risk-based standards and implementing regulations of the US Environmental Protection Agency (EPA) (i.e., 40 CFR 191 and 40 CFR 194, respectively) [4,5]. An FEP screening can be based on either a low-consequence or low-probability rationale. A low-probability rationale includes either (a) a qualitative rationale that the FEP is not credible or (b) a quantitative demonstration that the probability is less than 10-4 in 104 years. In this evaluation, a qualitative lowprobability rationale of not credible is used by demonstrating that bounding configurations of the waste are not critical. The demonstration of subcriticality is through quantitative calculations, but a probability of criticality is not evaluated. Rather, the rationale for this evaluation is that bounding configurations with an effective neutron multiplication factor (keff) well below the upper subcriticality limit (USL) make criticality incredible. Reference [2] documented a nuclear criticality assessment of the WIPP repository for disposal of dilute surplus plutonium materials using the Dilute and Dispose Approach and packaging in criticality control overpacks (CCOs). The CCO is the waste disposal container recently designed to allow for up to 380 fissile gram equivalent (FGE) 239 Pu per drum, which is a higher fissile loading than typical waste containers. The CCO consists of a criticality control container (CCC) positioned by upper and lower plywood spacers within a standard 55 gal drum. The CCC is used to establish a geometry control for fissile materials during transportation and WIPP emplacement operations. The current WIPP waste acceptance criteria for CCO payloads limit beryllium to less than or equal to 1% by weight of the waste contents and require the waste form to be non-machine compacted. Reference [2] considered two scenario progressions—room closure from salt creep, hereafter referred to as the reconfigured dry scenario, and flooding with brine, hereafter referred to as the reconfigured wet scenario. The subsequent drying out of the reconfigured wet scenarios was also considered. For all scenarios, subcriticality was maintained when 50 g of B 4 C (acting as a neutron absorber) per CCC was intermixed within the plutonium disposition waste form. The analysis used a waste form description that limits the amount of moderation that could be present within the waste form (i.e., it limits the amount of water and polyethylene that could be present based on planned processing conditions). This analysis to evaluate increased limits on the amount of moderation that could be present was performed as a companion to Reference [2] to address concerns associated with verifying moisture and/or plastic contents of waste materials following packaging of dilute surplus plutonium in the CCO. To that end, this analysis used the models and methods from Reference [2] to evaluate a more generic base waste form consisting of water and polyethylene that is more similar (and nearly identical) to the generic waste forms utilized in other models/analyses supporting the TRU Package Transporter Model II (TRUPACTII) safety analysis [6] (all are without moderation controls). The waste form in this analysis uses a base mixture of 75% water and 25% polyethylene, the total amount of which is varied to determine the optimum moderation to fissile material (H/Pu) ratio. The fissile loading is maintained at up to 380 FGE 239 Pu (modeled as PuO 2 ) per CCO with an additional 545 g of beryllium (to bound the 1% by weight contents restriction) and 50 g of B 4 C intermixed per CCO. The beryllium content (1% by weight) is based on the total allowed waste weight (this does not include packaging and container weights). Figures ES-1 and ES-2 display summary results, showing that with this model including 50 g of B 4 C per CCO, the system keff remains under 0.85 for all moderator amounts and provides a significant margin against post-closure criticality under postulated bounding conditions for compaction. Figure ES-1 compares an infinite model with a room model at the initial emplacement spacing and under full radial compaction. Full radial compaction places each CCC in direct contact and does not credit any anticipated spacing associated with current post-closure geomechanical modeling of the repository [7]. The effects of variations in the H/Pu ratio were evaluated by varying the amount of the water/polyethylene component of the waste model, with fissile loading maintained at 380 239 Pu FGE. Similarly, Figure ES-2 illustrates how various amounts of B 4 C per CCO influence k eff at different radial compactions, all at the H/Pu ratio of 200 (in the room array model). Therefore, while the results from Saylor 2017 [2] modeled more realistic process limits associated with packaging of dilute surplus plutonium, this analysis demonstrates that limits on moderation (plastic and water content) are not necessary to ensure subcriticality in the WIPP repository, provided the requisite B 4 C absorber is present.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

Initial Design of SABRE Fueled Molten Salt Experiment Irradiation Vehicle

Molten Salt Reactors (MSRs) are emerging as promising advanced reactor technologies, utilizing molten salts for both fuel and primary cooling. These reactors may offer advantages such as passive safety, enhanced economic viability, and efficient waste reprocessing while operating at high temperatures and low pressures, which leads to increased system efficiency and reduced mechanical stress on containment structures. However, challenges arise from the complex chemistry and high corrosion rates of fueled molten salts, alongside the volatility of certain fission products during irradiation. The accumulation of these fission products can alter the fuel salt chemistry, affecting corrosion potential and radioactive source terms. Therefore, understanding these phenomena under neutron irradiation is essential for future MSR designs. The SABRE (Salt and Actinide Burnup in a Reactor Environment) experiment, currently in conceptual design at Idaho National Laboratory (INL), aims to investigate these challenges by conducting a drop-in capsule experiment in the Advanced Test Reactor (ATR). The primary objectives include achieving a minimum burnup of 2 GWd/MTU and characterizing fission products while studying corrosion behavior in molten salt systems. This innovative experiment will leverage recent advancements in ATR capabilities, allowing for the safe irradiation of molten salts, thereby advancing the understanding and qualification of MSR technology for future nuclear power generation.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Methylmercury Speciation and Retention Evaluation to Support Saltstone Waste Acceptance Criteria

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Hierarchical Materials as Tailored Nuclear Waste Forms: A Perspective

This perspective focuses on the synthesis, characterization and modeling of three classes of hierarchical materials with potential for sequestering radionuclides: nanoparticles, porous frameworks and crystalline salt inclusion phases. The scientific impact of hierarchical structures and the development of the underlying crystal chemistry is discussed as laying the groundwork for the design, local structure control, and synthesis of new forms of matter with tailored properties. This requires development of the necessary scientific understanding of such complex structures through integrated synthesis, characterization, and modeling studies that can allow their purposeful creation and properties. The ultimate practical aim is to provide the means to create novel structure types that can simultaneously sequester multiple radionuclides. The result will lead to the creation of safe and efficient, long lasting waste forms for fission products and transuranic elements that are the products of nuclear materials processing waste streams. The generation of the scientific base for working toward that goal is presented.

hierarchal material, review article, nuclear waste↗

DOE OSTI2019

Screening of Additional Candidate Low-Temperature Glasses for Technetium Recovered from Hanford Effluent Management Facility Evaporator Bottoms

In this annex to a previously reported study, a series of laboratory screening tests was conducted to identify low-temperature glasses that have significant potential for use in disposition of technetium from a Hanford Tank Waste Treatment & Immobilization Plant (WTP) secondary liquid waste stream. Specifically, a glass waste form is sought that would be suitable for on-site disposition of technetium, salts, and other species in a Tc-rich stream recovered by ion exchange from the future Effluent Management Facility (EMF) evaporator concentrate. In the current study, several additional candidate glasses were identified, and testing of previous candidates was extended to include the Toxicity Characteristic Leaching Procedure (TCLP). In non-radioactive testing with rhenium surrogate, new formulations incorporated a simulated ion exchange eluate at temperatures below 750 °C to produce an amorphous, or mixed crystalline-amorphous, product. Incorporation of technetium (NaTcO 4 )-containing simulated eluate was evaluated for two of these glasses. Technetium was not retained well in these glasses, confounding the previous conclusion that Tc volatilization during vitrification can be reduced by appropriate selection of low-temperature glasses. Preliminary data for additional low-temperature glass compositions and recommendations for future work are provided. This report is a deliverable associated with requisition 55068-004-03 NFE-17-0020 "Candidate Low-T Glass Waste Forms for EMF Bottoms On-Site Disposition Alternative."

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DOE OSTI2017

Results of initial analyses of the salt (macro) batch 11 Tank 21H qualification samples

Savannah River National Laboratory (SRNL) analyzed samples from Tank 21H in support of qualification of Interim Salt Disposition Project (ISDP) Salt (Macro) Batch 11 for processing through the Actinide Removal Process (ARP) and the Modular Caustic-Side Solvent Extraction Unit (MCU). This document reports the initial results of the analyses of samples of Tank 21H. Analysis of the Tank 21H Salt (Macro) Batch 11 composite sample indicates that the material does not display any unusual characteristics or observations, such as floating solids, the presence of large amounts of solids, or unusual colors. Further sample results will be reported in a future document. This memo satisfies part of Deliverable 3 of the Technical Task Request (TTR).

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DOE OSTI2020

FY20 Update on Brine Availability Test in Salt. Revision 4

This report summarizes the 2020 fiscal year (FY20) status of the borehole heater test in salt funded by the US Department of Energy Office of Nuclear Energy (DOE-NE) Spent Fuel and Waste Science & Technology (SFWST) campaign. This report satisfies SFWST level-two milestone number M2SF-20SNO10303032. This report is an update of an August 2019 level-three milestone report to present the final as-built description of the test and the first phase of operational data (BATS la, January to March 2020) from the Brine Availability Test in Salt (BATS) field test.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Nitrate and Nitrite Incompatibility with Hydroxide Ions in Concentrated NaOH Solutions: Implications for Hydroxide and Gibbsite Reactivity in Alkaline Nuclear Waste

Electrolyte solutions in alkaline nuclear waste contain aluminate, hydroxide, nitrate and nitrite with sodium as the predominant counterion. The salts of these ions are highly soluble, so the liquids are highly concentrated. This study found that there is a substantial incompatibility between the hydroxide and nitrate and/or nitrate ions. This was determined by the observations that adding just one molal of NaNO2 or NaNO3 to saturated NaOH solution precipitation of 12 moles of NaOH·H2O salt, whereas the common ion effect would have expected only about 1 mole to precipitate. Further analysis indicates that the presence of nitrate and nitrite drastically increases the reactivity of sodium hydroxide ions in solution, which likely influences the reactivity of other hydroxide-mediated reactions. This enhanced reactivity is likely because it disrupts large Na+-OH- ion networks because nitrate and nitrite do not fit in those ion networks similarly to how some ions cannot substitute into a foreign crystal lattice. In contrast, the aluminate ion did not have the same large incompatibility with hydroxide.

Nitrite, Nitrate, Aluminate, NaOH*H2O, Ion-aggrega↗

DOE OSTI2018

Regional Geologic Evaluations for Disposal of HLW and SNF: Alluvial Basins of the Basin and Range Province

The DOE Spent Fuel and Waste Technology (SFWT) R&D Campaign supports research on crystalline rock, shale (argillite) and salt as potential host rocks for disposal of HLW and SNF in a mined geologic repository. Modeling involving these host rocks for the generic GDSA reference cases has assumed saturated conditions at repository depths. In order to test two-phase flow capabilities of the GDSA flow and transport models, the SFWT Campaign is developing an unsaturated zone reference case. Alluvial basin-fill deposits within the Basin and Range province have unsaturated zones that can be as thick as 600 meters, although the thickness is typically much less. Basin-fill alluvium has been considered as a potential medium for HLW disposal in the US since the 1950s (NAS,1957; Smyth et al., 1979; Winograd, 1981). Attributes considered favorable for HLW disposal in alluvium in the Basin and Range province include an arid environment, low water flux , high sorption capacity (from clay s and zeolites), and ease of construction and stability of repository openings (Hardin et al., 2012). Hardin et al. (2012) also address the thermal issues inherent in the design of a HLW repository in alluvium with low thermal conductivity, which are not discussed in this report.

58 GEOSCIENCES↗

DOE OSTI2019

Impacts of Fast-Spectrum Molten Salt Reactor Characteristics on Fuel Cycle Performance

The System Analysis and Integration Campaign nuclear fuel cycle evaluation and screening study was conducted to outline the potential benefits and challenges of nuclear fuel cycle options. The study identified continuous recycle in fast critical reactors as a common characteristic shared among some of the most promising future nuclear fuel cycles. This study was technology agnostic to focus on the underpinning physics that drives fuel cycle outcomes. This process used an analysis example to generate a physics-based understanding of the fuel cycle, and it yielded performance metrics over ranges rather than comparing absolute numbers. While no fast-spectrum molten salt reactor (MSR) has ever been built, several concepts exist using different fuels and carrier salts, many of which target high-burnup objectives, such as lightwater reactor nuclear waste consumption. Putting the resulting fuel cycle impacts of these reactor choices (e.g., fuel and neutron spectrum) into perspective is informative for our current understanding of a given fuel cycle’s performance, and it also demonstrates the applicability of MSRs to different fuel cycles. It has been established that in a fuel cycle transition, performance of a given fast-spectrum MSR depends in part on the initial fuel loading requirements and processing methods.

Betzler, Benjamin↗

DOE OSTI2018

Effects of Ionic Strength, Salt, and pH on Aggregation of Boehmite Nanocrystals: Tumbler Small-Angle Neutron and X-ray Scattering and Imaging Analysis

The U.S. government currently spends significant resources annually managing the legacies of the Cold War, including 300 million liters of highly radioactive wastes stored in hundreds of tanks at the Hanford Site (WA) and the Savannah River Site (SC). The materials in these tanks consist of highly radioactive slurries and sludges at very high pH and salt concentrations. The solid particles primarily consist of aluminum hydroxides and oxyhydroxides (gibbsite and boehmite), although many other materials are present. These form complex aggregates that dramatically affect the rheology of the solutions and, therefore, efforts to recover and treat these wastes. In this paper we have used a combination of transmission and cryo-transmission electron microscopy, dynamic light scattering, and X-ray and neutron small- and ultra-small angle scattering to study the aggregation in controlled environments of synthetic nano-boehmite particles at pH 9 (approximately the point of zero charge) and 12, and sodium-nitrate and calcium-nitrate concentrations up to 1 molal. While the initial particles form individual rhombohedral platelets, once placed in solution they quickly form well-bonded stacks, primary aggregates, up to ~ 150 nm long. These are more prevalent at pH=12. Addition of calcium-nitrate or sodium-nitrate has a similar effect as lowering pH, but approximately one hundred times less calcium than sodium is needed to observe this effect. These aggregates have fractal dimension between 2.5 and 2.6 that are relatively unaffected by salt concentration for calcium-nitrate at high pH. Larger aggregates (> ~ 400 nm) are also formed, but their size distributions are discrete rather than continuous. The fractal dimensions of these aggregates are strongly pH dependent but only become dependent on solute at high salt concentrations.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Project Plan: Salt in Situ Heater Test.

This project plan gives a high-level description of the US Department of Energy Office of Nuclear Energy (DOE-NE) Spent Fuel and Waste Disposition (SFWD) campaign in situ borehole heater test project being planned for the Waste Isolation Pilot Plant (WIPP) site This plan provides an overview of the schedule and responsibilities of the parties involved. This project is a collaborative effort by Sandia, Los Alamos, and Lawrence Berkeley National Laboratories to execute a series of small-diameter borehole heater tests in salt for the DOE-NE SFWD campaign. Design of a heater test in salt at WIPP has evolved over several years. The current design was completed in fiscal year 2017 (FY17), an equipment shakedown experiment is underway in April FY18, and the test implementation will begin in summer of FY18. The project comprises a suite of modular tests, which consist of a group of nearby boreholes in the wall of drifts at WIPP. Each test is centered around a packer-isolated heated borehole (5" diameter) containing equipment for water-vapor collection and brine sampling, surrounded by smaller-diameter (2" diameter) satellite observation boreholes. Observation boreholes will contain temperature sensors, tracer release points, electrical resistivity tomography (ERT) sensors, fiber optic sensing, and acoustic emission (AE) measurements, and sonic velocity sources and sensors. These satellite boreholes will also be used for plugging/sealing tests. The first two tests to be implemented will have the packer-isolated borehole heated to 120°C, with one observation borehole used to monitor changes. Follow-on tests will be designed using information gathered from the first two tests, will be conducted at other temperatures, will use multiple observation boreholes, and may include other measurement types and test designs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Rare Earth Electrochemical Property Measurements and Phase Diagram Development in a Complex Molten Salt Mixture for Molten Salt Recycle

Pyroprocessing is a promising alternative for the reprocessing of used nuclear fuel (UNF) that uses electrochemical methods. Compared to the hydrometallurgical reprocessing method, pyroprocessing has many advantages such as reduced volume of radioactive waste, simple waste processing, ability to treat refractory material, and compatibility with fast reactor fuel recycle. The key steps of the process are the electro-refining of the spent metallic fuel in the LiCl-KCl eutectic salt, which can be integrated with an electrolytic reduction step for the reprocessing of spent oxide fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Volatile species of technetium and rhenium during waste vitrification

Volatile loss of technetium (Tc) during vitrification of low-activity wastes is a technical challenge for treating and immobilizing the large volumes of radioactive and hazardous wastes stored at the U.S. Department of Energy's Hanford Site. There are various research efforts being pursued to develop technologies that can be implemented for cost effective management of Tc, including studies to understand the behavior of Tc during vitrification, with the goal of eventually increasing Tc retention in glass. Furthermore, one of these studies has focused on identifying the form or species of Tc and Re (surrogate for Tc) that evolve during the waste-to-glass conversion process. This information is important for understanding the mechanism of Tc volatilization. In this paper, available information collected from the literature is critically evaluated to clarify the volatile species of Tc and Re and, more specifically, whether they volatilize as alkali pertechnetate and perrhenate or as technetium and rhenium oxides after decomposition of alkali pertechnetate and perrhenate. The evaluated data ranged from mass spectrometric identification of species volatilized from pure and binary alkali pertechnetate and perrhenate salts to structural and chemical analyses of volatilized materials during crucible melting and scaled melter processing of simulated wastes.

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